The following relates to the nuclear power generation arts, nuclear reactor safety arts, nuclear reactor emergency core cooling (ECC) arts, and related arts.
In a loss of coolant accident (LOCA) or other event in which the pressure vessel of a nuclear power plant is depressurized, the nuclear reactor core is to be kept immersed in water so as to provide for removal of decay heat and to prevent exposure of the fuel rods to air which can lead to chemical reactions and release of airborne radioactivity. The system which provides this water injection is referred to as the emergency core cooling (ECC) system. In a typical arrangement, high pressure water injection tanks employing passive nitrogen pressurization operate during the initial reactor depressurization to force pressurized water into the reactor pressure vessel to ensure the reactor core remains immersed in water during the depressurization process. After the reactor is depressurized, a refueling water storage tank (RWST) located with the nuclear reactor at an elevated position inside the radiological containment is drained into the reactor pressure vessel. This flow of water is a passive, gravity-driven process, and with a suitably sized RWST can be continued without any supply of external power or water for a design period of hours or days. In the case of a LOCA, steam or two-phase steam/water mixture is discharged from the reactor pressure vessel into the surrounding radiological containment, and this water condenses and collects in a sump of the radiological containment. A sump pump may be provided to recover this water into the RWST. Closed heat exchange loops may also be provided to form steam cycle heat transfer from the pressure vessel to the RWST or to other heat dissipation systems. These systems allow the reactor to be shut down in the case of a LOCA or other sudden depressurization event with little or no damage to the nuclear reactor core.
Additionally, nuclear regulatory rules, e.g. promulgated by the Nuclear Regulatory Commission (NRC) in the United States, typically require that the nuclear power plant incorporate safety systems to protect against the possibility of loss of reactor core cooling to an extent resulting in some melting of the reactor core. In conventional light water reactors employing 235U as the fissile radioisotope, the nuclear fuel is typically in the form of uranium dioxide (UO2) which has a melting point of around 3100K. At this temperature, the nuclear fuel assemblies and the steel structures around the nuclear reactor core melt and form a molten mass of material that is referred to in the art as “corium”.
In an ex-vessel retention approach, the nuclear power plant is designed for a contingency in which the corium relocates to the bottom of the reactor vessel, melts through the bottom of the reactor pressure vessel, and collects on the containment floor below the reactor vessel. In the ex-vessel retention approach, the floor below the reactor vessel is lined with high temperature zirconia thermal insulation tiles to minimize interaction with the underlying concrete forming the floor of the sump. In most scenarios, the sump is filled with water, e.g. condensed steam released by a LOCA. The molten corium spreads out over the sump floor, expanding its surface area and rapidly cooling. If the corium melts through the pressure vessel rapidly, then the potential exists for a steam explosion when the corium comes into rapid contact with water in the sump of the radiological containment structure. This steam explosion can be a challenge to the structural design of the containment structure.
In an in-vessel retention approach, a flow channel is provided between the reactor pressure vessel and the surrounding insulation panels. This flow channel is designed to allow water flow around the lower vessel when the reactor cavity (i.e. radiological containment sump) is flooded. The goal is to cool the vessel sufficiently to prevent the corium inside the vessel from melting though the vessel shell, thereby mitigating the likelihood of an ex-vessel steam explosion scenario. However, in-vessel retention potentially allows long-term exposure of the molten corium to air which can lead to release of airborne fission products. The survival of the lower vessel is also dependent upon the amount of molten material, its melt temperature, and physical configuration of molten layers on the bottom of the vessel. In view of this, nuclear regulatory rules typically require that a nuclear power plant designed for in-vessel retention additionally be designed to account for an ex-vessel corium relocation contingency, including the possibility of a steam explosion due to rapid melting of the pressure vessel and corium contact with water collected in the vessel cavity.